Mirco Große

• Senior Researcher at Karlsruhe Institute of Technology

When hydrogen penetrates into cladding tubes during operation of nuclear power plants, it may affect their mechanical stability. During the dry storage of spent nuclear fuel, the initially dissolved hydrogen precipitates as hydrides in the metallic matrix. This precipitation process induces a certain stress in the zirconium structure of the claddin...

A long-term bundle tests lasting 250 days were conducted at the LICAS facility in the Karlsruhe Institute of Technology as part of the SPIZWURZ project. The pressurized fuel rod simulators with commercial prehydrided non-irradiated claddings were tested under conditions close to the ones of dry storage of spent nuclear fuel. This paper compares the...

Accident tolerant fuel (ATF) cladding is a new type of nuclear fuel cladding designed to improve the safety and performance of nuclear reactors. In this paper, the kinetics and degradation mechanisms during high-temperature oxidation in steam of the three most promising ATF cladding materials, i.e., chromium-coated zirconium alloys, FeCrAl alloys,...

Neutrons interact with the magnetic moment of the atomic shell of an atom, as is common for X-rays, but mainly they interact directly with the nucleus. Therefore, the atomic number and the related number of electrons does not play a role in the strength of an interaction. Instead, hydrogen that is nearly invisible for X-rays has a higher attenuatio...

During the dry storage of spent nuclear fuel, the initially dissolved hydrogen precipitates as hydrides in the metallic matrix. The orientation of hydrides influences the crack propagation and depends mainly on the hoop stress. In the frame of the SPIZWURZ project, the reorientation of zirconium hydrides in cladding tubes is being investigated unde...

During operation in light-water reactors, Zircaloy cladding tubes take up hydrogen that precipitates under specific temperature–pressure conditions after operation in the form of zirconium hydrides. These zirconium hydrides have a detrimental effect on the mechanical stability of the cladding tubes. The conditions for their formation and their orie...

Zirconium (Zr) alloys are widely used in nuclear power plants as fuel cladding and are susceptible to hydrogen (H) degradation. For long operational service, Zr-based components can suffer a mechanism known as Delayed Hydride Cracking (DHC) associated to an increase of the crack propagation velocity by the re-orientation and precipitation of Zr hyd...

In order to in-situ quantify the hydrogen diffusion in metals or more precisely in zircaloy cladding tubes with the influence of an applied stress field, a new device was constructed in cooperation with the company ZwickRoell - the transportable INCHAMEL facility. It is a modification of ZwickRoell’s Kappa Mini 1 kN tensile testing machine. The fac...

Delayed Hydride Cracking (DHC) is a failure mechanism that occurs in Zr alloys under certain conditions of hydrogen concentration, temperature and stress gradient. In service, hydrogen produced by corrosion reaction can be incorporated in Zr alloys and if the solid solubility is exceeded, hydrogen precipitates as zirconium hydride. The presence of...

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. A type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal...

The annealing behavior of the pre-oxidized Cr-coated Zry-4 at 1200°C in argon is systematically investigated. A Cr2O3 scale with a thickness of ∼9 μm formed on the sample surface after pre-oxidation of the Cr-coated Zry-4 in steam at 1200°C for 30 min. During annealing in inert atmosphere, the thickness of the Cr2O3 scale decreases with the increas...

The experiment QUENCH-20 with BWR geometry simulation bundle was successfully conducted at KIT on 9th October 2019 in the framework of the international SAFEST project. The test bundle mock-up represented one quarter of a BWR fuel assembly with 24 electrically heated fuel rod simulators and two B4C control blades. The rod simulators were filled wit...

The CODEX-AIT-3 experiment simulated an in-vessel air ingress scenario after failure of bottom head of a nuclear reactor. 1600 °C maximum temperature was reached with the electrically heated bundle and slow cool-down was applied at the end of the test to demonstrate the bundle state before quench. Limited air and steam flow rates were intentionally...

The QUENCH-19 bundle experiment was the worldwide first bundle test simulating severe accident conditions with ATF cladding materials. It was conducted with FeCrAl(Y) cladding tubes (alloy B136Y3, developed by the Oak Ridge National Lab, USA) and 4 Kanthal AF spacer grids as well as 7 Kanthal APM corner rods and Kanthal APM shroud was conducted at...

Single-rod oxidation and quench experiments at very high temperatures in steam atmosphere were conducted with advanced, nuclear grade SiCf/SiC CMC cladding tube segments. A transient experiment was performed until severe local degradation of the sample at maximum temperature of approximately 1845 °C. The degradation was caused by complete consumpti...

The oxidation mechanism and kinetics of two nuclear-grade FeCrAl alloys were investigated in steam up to 1500°C by transient and isothermal oxidation tests. The slow α-alumina formation kinetics well matched only for the temperature range from 1000°C to 1300°C. Below 1000°C, formation of transient alumina caused faster kinetics. In addition, an exc...

The isothermal oxidation behavior of the Cr-coated Zircaloy-4 at 1200 °C in steam is comprehensively investigated. Oxidation kinetics transition is observed when outward diffusion of Zr and precipitation of ZrO2 along the Cr coating grain boundaries are reaching the Cr/Cr2O3 interface. The outward diffused Zr reduces the Cr2O3 scale to Cr. Pores fo...

Elemental Cr/C/Al multilayers (stoichiometric ratio: 2:1:1) with and without a Cr overlayer have been synthesized on Zircaloy-4 substrates by magnetron sputtering. The effects of annealing temperatures (400 and 550 °C) on phase/microstructure formation, mechanical properties, and oxidation/corrosion performance have been comparatively studied. Anne...

Zirconium alloys in nuclear power plants operate in high-pressure water at temperatures between 250°C-350°C. Hydrogen (or deuterium) ingress due to waterside corrosion and if the solubility is exceeded H precipitates as a brittle hydride phase. Degradation mechanisms involve the accumulation of these brittle hydrides at cold spots or crack tips, as...

Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on proto...

The Fukushima-Daiichi accident revealed that the zirconium fuel claddings have the significant safety risk of hydrogen detonation due to the strong oxidation and hydrogen release during the design basis accidents (DBA) and beyond design basis accidents (BDBA). Therefore, research and development of accident tolerant fuel (ATF) concepts that aim to...

The transient oxidation behavior of magnetron-sputtered chromium-coated Zircaloy-4 was studied in steam up to 1600°C, and the microstructural evolution of the coating-substrate system after oxidation was investigated. Coating failure and corresponding rapid oxidation of coating and substrate occurred at 1300–1400°C. It was mainly caused by the thic...

The effects of atmosphere (oxygen and steam) and reactive element addition (Zr and Y) on high-temperature oxidation behavior of Al0.5CrFeNiMn and Al0.5CrFeNiCo high-entropy alloys at 1000°C and 1200°C were studied. The Al0.5CrFeNiMn alloy displayed different oxidation mechanisms in oxygen and steam owing to differing oxygen partial pressures and pr...

Description Featuring 37 peer-reviewed and award-winning papers from industry experts that were presented at this 2019 symposium held in Manchester, United Kingdom. Topics covered include Sponsored by ASTM International Committee B10 on Reactive and Refractory Metals and Alloys and its subcommittee Zirconium and Hafnium.

Featured Application The investigations provide information about important parameters needed for the modelling of design bases, and beyond design bases, nuclear accidents and processes occurring during the long-term dry storage of spent nuclear fuel. Abstract In situ neutron radiography experiments can provide information about diffusive processe...

Alumina-forming MAX phase coatings reveal great potential for accident tolerant fuel (ATF) cladding applications due to their favorable physical and mechanical properties and excellent high-temperature oxidation resistance. The feasibility of the Cr2AlC MAX phase as protective coating on zirconium-alloy fuel claddings was explored focusing on its h...

【Please note】The full-text can be download for free until 25.03.2021 via this link: https://authors.elsevier.com/a/1cWWt54hEI9jd ....................................................................................................................................... Abstract: The Cr-Zr interaction of two types of Cr coated Zr alloy accident toleran...

The investigations in this paper deal with the Cr-Ni alloy. The material has been recently proposed as a potential ATF concept, primarily due to its behaviour under high-temperature oxidation. A set of experiments to determine the melting point and describe the oxidation kinetics of the Cr-Ni alloy were performed in Karlsruhe Institute of Technolog...

The paper presents a summary of the round robin test activity organized within the IAEA ACTOF project. The test conditions and sample matrix were finalized during fall 2017, production and transportation of samples started in Q1 2018 and the tests have been performed during the past 14 months. Two fundamental experimental tests related to normal op...

The QUENCH-LOCA bundle test series was launched to investigate the influence of the secondary hydriding phenomena on the applicability of the cladding embrittlement criteria. Seven out-of-pile bundle tests with different zirconium alloy-based cladding materials were performed according to a temperature/time-scenario typical for a LBLOCA in German P...

Oxidation tests of alumina-forming austenitic (AFA) alloys and high entropy alloys (HEA) have been performed in steam at 1200 °C. AFA alloys with composition formula Fe-3Al-16Cr-(19-21)Ni (wt.%) and HEA with formula Al(7.9-8.9)Cr(21.4-23.2)Ni(34.3-35)Febal (at.%) show low oxidation rate due to the formation of α-Al2O3 scale. By adding yttrium to AF...

The fuel rod claddings in nuclear light water reactors are made of zirconium alloys. Corrosion of these alloys during operation and in particular high temperature oxidation during nuclear accidents results in the production of free hydrogen. The cladding can absorb this hydrogen. It affects the mechanical properties of the cladding material. Hydrog...

The oxidation performance and quench behavior of cold spraying Cr coated Zircaloy-4 cladding tubes were investigated from 1100°C up to 1500°C in steam. The coated samples displayed significantly improved oxidation resistance, good thermal shock resistance and high post-quench ductility during oxidation at 1100°C and 1200°C for 1 hour accompanied by...

The aim of the IL TROVATORE project is the development, characterization and testing of potential materials for nuclear fuel rod claddings with enhanced accident tolerance. One main task of the ongoing investigations are testing of the cladding/coolant interactions under operational as well as accident conditions. A wide range of materials was alre...

Experiment QUENCH-20 with BWR geometry simulation bundle was successfully conducted at KIT on 9th October 2019. This test was performed in the framework of international access SAFEST infrastructure with the users from Swedish Radiation Safety Authority (SSM) in cooperation with Westinghouse Sweden, GRS and KTH. The test objective was the investiga...

In the framework of the QUENCH program at KIT, over the past 22 years, 21 bundle tests were performed under severe accident conditions with different cladding materials. Additionally, 7 QUENCH-LOCA bundle tests with fresh and pre-hydrogenated different cladding materials (Zry-4, M5®, opt. ZIRLO™) were performed according to a temperature/time-scena...

The QUENCH-19 experiment with the test bundle including twenty four electrically heated fuel rod simulators with FeCrAl(Y) claddings was conducted at KIT on 29th August 2018. Other bundle parts were four FeCrAl(Y) spacer grids as well as eight KANTHAL APM corner rods; the bundle was surrounded by KANTHAL APM shroud. The test objective was the compa...

A preliminary analysis of the bundle reflood experiment QUENCH-18 is performed with the SCDAPSim/Mod3.5/da code containing the PSI-developed model for oxidation in the presence of air. The simulation follows on from pre-test planning and prediction calculations using the same code and input model. The starting point for the post-test calculations d...

Hydrogen distribution in zirconium nuclear fuel claddings can often be non-uniform because of the high mobility of hydrogen interstitial atoms, raising the risk to the fuel rod integrity. Therefore, the assessment of hydrogen migration behavior is of great importance when considering its potential detrimental effects during spent fuel storage. Here...

Absorbed hydrogen degrades the mechanical properties of zirconium alloys used for nuclear fuel claddings. Not only the total amount of hydrogen absorbed in the cladding tube but also the zirconium hydride orientation and its distribution influence the toughness of the material. For instance, the so-called delayed hydride cracking is caused by the d...

In the framework of the SARNET-2 European program the QUENCH-DEBRIS test was conducted as the 17th severe accident test using the QUENCH facility at KIT to investigate the formation and coolability of a prototypic debris bed. The test scenario was defined by pre-test calculations using the MELCOR code. The test bundle with a length of about 2 m con...

FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500{\deg}C was examined. In case of tr...

FeCrAl alloys are proposed and being intensively investigated as alternative accident tolerant fuel (ATF) cladding for nuclear fission application. Herein, the influence of major alloy elements (Cr and Al), reactive element effect and heating schedules on the oxidation behavior of FeCrAl alloys in steam up to 1500 °C was examined. In case of transi...

RESUMEN Las aleaciones base circonio (Zr) se utilizan en la fabricación de las vainas de combustible y en componentes estructurales de los reactores de potencia por su baja absorción neutrónica. Estas aleaciones son susceptibles al daño por hidrógeno (H), debido a que la concentración del mismo puede incrementarse localmente en respuesta a gradient...

The QUENCH-L5 experiment was performed in the framework of the QUENCH-LOCA test series. The overall objective of this bundle test series is the investigation of ballooning, burst, degree of oxidation and secondary hydrogen uptake of the cladding under representative design-basis accident conditions and their influence on the mechanical properties....

The QUENCH-L4 experiment was performed in the framework of the QUENCH-LOCA test series. The overall objective of this bundle test series is the investigation of ballooning, burst, degree of oxidation and secondary hydrogen uptake of the cladding under representative design-basis accident conditions and their influence on the mechanical properties....

The QUENCH-L3 experiment was performed in the framework of the QUENCH-LOCA test series. The overall objective of this bundle test series is the investigation of ballooning, burst, degree of oxidation and secondary hydrogen uptake of the cladding under representative design-basis accident conditions and their influence on the mechanical properties....

The QUENCH-L2 bundle experiment was performed in the framework of the QUENCH-LOCA test series. As-received M5® tubes with an outside diameter of 10.75 mm were used as claddings of 21 electrical heated rods. Similar to the previous QUENCH-L1 test, the fuel rod simulators were separately internally pressurized with krypton to 55 bar. Bundle configur...

Alumina-forming MAX phase ternary carbides are being considered as protective coatings on zirconium alloys as accident tolerant fuel (ATF) cladding because of their resistivity against high-temperature steam oxidation during accident scenarios. This study attempted to synthesize three types of Al-containing MAX phase carbides (Ti2AlC, Cr2AlC and Zr...

Stacked couples of silicon carbide and Zircaloy-4 discs were annealed for 1 h in nominal inert atmosphere (6N Ar) at temperatures of 1200, 1400, 1500, 1550, 1575 and 1600°C. Strong interactions between silicon carbide and Zircaloy-4 occurred at temperatures of 1500°C and above. The width of the influenced zone exceeds 1 mm at 1550°C. A pronounced l...

Zr-based alloys are used in nuclear power plants because of a unique combination of very low neutron absorption and excellent mechanical properties and corrosion resistance at operating conditions. However, Hydrogen (H) or Deuterium ingress due to waterside corrosion during operation can embrittle these materials. In particular, Zr alloys are affec...

In the framework of nuclear safety research the high temperature reaction behavior of the classical pressurized water reactor cladding material Zircaloy-4 in nitrogen containing wet atmospheres was investigated by means of in-situ neutron radiography. The paper describes experimental details and gives an overview of the results. The dependence of t...

The QUENCH-L1 bundle experiment with Zircaloy-4 cladding tubes was defined as reference test for the QUENCH-LOCA test series. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design based accident conditions, as well as to check the embrittl...

Hydrogen uptake by nuclear fuel claddings during normal operation as well as loss of coolant during design basis and severe accidents beyond design basis has a high safety relevance because hydrogen degrade the mechanical properties of the zirconium alloys applied as cladding material. Currently, claddings with enhanced accident tolerance are under...

Oxidation of single-phase and dense Ti2AlC coatings with or without a 500 nm TiC diffusion barrier deposited on Zircaloy-4 by annealing of nanoscale multilayer stacks between 800 °C and 1200 °C in high-temperature steam was investigated. Coatings without TiC barrier formed a duplex scale: outer θ-Al2O3 rich layer mixed with TiO2 and inner porous Ti...

Description Get 43 papers from global researchers on advancements in zirconium technology in the nuclear industry. Learn more about the range of challenges for the nuclear community and how they are being addressed in top research presented by industry associates, national laboratories, and universities. This publication includes three papers from...

Description Get 43 papers from global researchers on advancements in zirconium technology in the nuclear industry. Learn more about the range of challenges for the nuclear community and how they are being addressed in top research presented by industry associates, national laboratories, and universities. This publication includes three papers from...

Current methodical developments improve the spatial resolution of neutron imaging facilities. Objects with dimensions down to several microns should be detectable. However, the minimum object size detectable depends not only on the facility hardware like detector resolution or neutron optics, but also on the attenuation contrast. In this paper the...

High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% ni...

A large-scale bundle experiment with silver-indium-cadmium absorber rod in the QUENCH facility as well as a series of small-scale single absorber rod tests with prototypical materials have been conducted. A variety of parameters like temperature history, initial contact between cladding and guide tube and possibility of inner oxidation of the guide...

The oxidation behavior of bulk Ti2AlC ceramic in steam has been investigated in the temperature range of 1400 °C–1600 °C. The oxidation kinetics followed a sub-parabolic law at the early stage of oxidation, then transferred to a linear law beyond 18 h at 1400 °C, and obeyed a linear law during the whole exposure up to 24 h at 1500 °C. At the initia...

Neutron imaging methods are appropriate to investigate hydrogen distributions in several metallic systems. The large total neutron cross section of hydrogen compared to those of elements or isotopes, respectively, in usual structural materials like steels or zirconium alloys allows the detection even of small amounts of hydrogen in such materials....

The oxidation behavior of two commercial FeCrAl alloys, Kanthal APM and D, with high Cr content (20.5 wt.% Cr, 4.8-5.8 wt.% Al) was investigated under well-defined heating schedules at 1300-1500°C in steam. In isothermal tests, both alloys melted at 1500°C, rapidly and completely oxidized at 1400°C, and formed protective alumina scales at 1300°C. R...

The hydrogen concentration and distribution at both sides of the burst opening of cladding tubes used in three QUENCH-LOCA simulation bundle experiments were investigated by means of neutron radiography and tomography. The quantitative correlation between the total macroscopic neutron cross-section and the atomic number density ratio between hydrog...

High material penetration by neutrons allows for experiments using sophisticated sample environments providing complex conditions. Thus, neutron imaging holds potential for performing in situ nondestructive measurements on large samples or even full technological systems, which are not possible with any other technique. This paper presents a new sa...

The most important accident management measure to terminate a severe accident transient in LWR is the injection of water to cool the uncovered degraded core. In order to detailed investigation of the reflood effect on bundle degradation the QUENCH program was initiated in 1996 followed-up the CORA bundle tests and is still ongoing. So far, 17 integ...

The QUENCH-L0 experiment was defined as commissioning test for the new QUENCH-LOCA test series. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design basis accident conditions as well as detailed post-test investigation of cladding mechani...

The oxidation of zirconium cladding alloys used in nuclear reactors was investigated under the conditions of loss-of-coolant and severe accidents, i.e., at temperatures between 600 and 1600°C and in various atmospheres. The kinetics were parabolic or sub-parabolic as long as the superficially formed oxide scale remained intact and protective. More...

LOCA simulation tests were performed in the QUENCH facility of KIT on fuel rod bundle scale. The first two tests using Zircaloy-4 claddings, out of a series of six tests with different cladding alloys, were already performed. The test conditions and results are described. The secondary hydrogenation of the Zircaloy-4 cladding tubes was investigated...

The objective of this paper is to summarize the results of the latest observations performed at Paul Scherrer Institut on irradiated fuel claddings, to characterize their corrosion and hydrogen-uptake behavior. Two categories of studies have been performed. (1) A series of destructive tests were achieved on the fuel rods irradiated in a boiling-wat...

Over the past 20 years, integral fuel bundle experiments performed at IRSN Cadarache, France (Phébus-SFD and Phébus FP – fission heated) and at Karlsruhe Institute of Technology, Germany (CORA and QUENCH – electrically heated), accompanied by separate-effect tests, have provided a wealth of detailed information on core degradation phenomena that oc...

Two out-of-pile bundle tests, QUENCH-L0 and QUENCH-L1, were performed recently at Karlsruhe Institute of Technology (KIT) in the framework of the QUENCH-LOCA program devoted to the investigation of the so-called secondary hydriding of the cladding. The overall objective of this bundle test series is the investigation of ballooning, burst and second...

In the framework of the SARNET-2 European program the QUENCH-DEBRIS test was conducted as the 17th severe accident test using the QUENCH facility at KIT to investigate the formation and coolability of a prototypic debris bed. The test scenario was defined by pre-test calculations using the MELCOR code. The test bundle with length of about 2 m conta...

The QUENCH-LOCA project on out-of-pile bundle tests under conditions of a loss of coolant reactor accident is integral part of the Nuclear Safety Program at the Karlsruhe Institute for Technology. The overall objective of this project is the investigation of ballooning, burst and secondary hydrogen uptake of the fuel cladding tubes under design bas...

In the framework of the IAEA Coordinated Research Project "Development, Characterization and Testing of Materials of Relevance for Nuclear Energy Sector Using Neutron Beams" 19 participants from 18 countries work together to study structural materials like steels and zirconium alloys by means of neutron diffraction, small angle neutron scattering,...

In the framework of the KIT QUENCH-LOCA program three loss of coolant accident simulation tests at fuel rod bundle scale were performed so far. From the results of the posttest examinations and of model calculations parameters influencing the secondary hydriding were identified. In the paper the effects of temperature, gap width between pellets and...

In the framework of the post-test examinations of the large-scale LOCA simulation tests at the fuel rod bundle scale, the hydrogen distributions in specimens prepared from the QUENCH-L0 and -L1 tests were studied by means of neutron radiography and tomography. In order to determine quantitative hydrogen concentrations, both, neutron radiography and...

The out-of-pile bundle experiment QUENCH-16 on air ingress was conducted in the electrically heated 21-rod QUENCH facility at KIT in July 2011. The oxidation of the Zircaloy-4 claddings in air following a limited pre-oxidation in steam was examined. Three contributors for intensive hydrogen production during the final reflood were identified: 1) re...

New out-of-pile QUENCH-LOCA bundle tests are being performed in the QUENCH facility within the Nuclear Safety Program of KIT. The overall objective of this bundle test series is the investigation of ballooning, burst and secondary hydrogen uptake of the cladding under representative design basis accident conditions as well as detailed post-test inv...

The hydrogen uptake and redistribution in Zircaloy-4 specimens applied to loss of coolant accident (LOCA) simulation experiments and in mechanical pre-loaded samples were investigated by means of ex-situ and in-situ neutron imaging. The results of these investigations were compared with results from mechanical tests. Hydrogen absorption may have a...

Hydrogen is one of the prominent life limiting factors for fuel cladding tubes. Although its diffusion behavior is well known for concentration and temperature gradient, few data exist to account for a stress driven diffusion. The problem in assessing such a behavior comes from the hydrides formation that depends on hydrogen concentration, temperat...

Neutron radiography was applied for investigations of nuclear fuel cladding and control rod behaviour during steam oxidation at temperatures between 1123 and 1673 K under severe nuclear accident conditions. This article gives an overview of these investigations. At KIT, loss of coolant and severe nuclear accidents were experimentally simulated. Pos...

The hydrogen diffusion from the gas phase into Zircaloy-4 solid cylinders was investigated at various temperatures between 823 and 1473 K. Diffusion coefficients were fitted using axial hydrogen distributions for all temperatures investigated. The activation energy of the hydrogen diffusion was calculated for temperatures at which the bcc β phase i...

This chapter introduces gas-solid and solid-solid reactions of reactor materials with the reactor atmosphere and with each other at high temperatures. The chapter focuses on the reaction of fuel cladding materials in different oxidising atmospheres. An overview of oxidation, nitriding, hydrogenation, carburisation and decarburisation of reactor str...